A conventional boiling water reactor (BWR) includes a pressure vessel containing a nuclear reactor core above which are disposed in turn conventional steam separators and dryers. The vessel is filled with a cooling and moderating fluid such as water to a predetermined normal water level located generally near the middle of the steam separators. The core boils the water for generating a steam-water mixture which rises upwardly into the steam separators, which remove some of the water therefrom, with additional water being further removed therefrom from the steam dryers positioned above the steam separators. The dried steam is conventionally discharged from the vessel to a conventional steam turbine, for example, which powers an electrical generator for generating electrical power provided to an electrical utility grid.
A typical BWR is controlled by a plurality of control rods which extend downwardly from the core through conventional guide tubes extending from the bottom of the core to the lower head of the pressure vessel which defines therebetween a lower plenum. Extending downwardly from the lower head are a plurality of conventional control rod drives (CRDs) which are effective for selectively inserting the control rods upwardly into the core for reducing reactivity therein, and for selectively withdrawing the control rod downwardly from the core for increasing reactivity therein. Accurate intermediate positions of the control rods may be obtained by using a conventional drive screw which is selectively rotated in opposite directions by a conventional stepper motor to selectively translate upwardly and downwardly a ball nut threadingly engaged therewith. An elongate piston rests on the ball nut and is coupled to a respective control rod for raising and lowering the control rod as the ball nut is correspondingly translated. In order to obtain relatively instantaneous insertion of the control rods during a SCRAM operation, a pressurized fluid such as water is conventionally channeled through the CRD for lifting the piston and in turn lifting the control rod independently of the ball nut.
In order to fully withdraw the control rods from below the core, the guide tubes extending between the core and the vessel lower head must have a vertical height approximately equal to the length of the control rods. The height of the core also has a vertical height approximately equal to the length of the control rods so that the control rods may be fully inserted therein. The conventional steam separators additionally require a suitable vertical height for effectively separating water from the steam-water mixture. And additional vertical height is required for the steam dryer disposed above the steam separators.
Accordingly, the overall height of the pressure vessel must be suitable for containing these several components and for allowing the effective functioning thereof. A typical pressure vessel for a BWR sized for generating steam to power a turbine-generator for providing electrical power to the electrical utility grid is about 21 meters tall, with the reactor generating on the order of about 1,000 megawatts electric (MWe) and higher. Such a large pressure vessel, which is typically made from steel, has a correspondingly high weight requiring large cranes for the assembly thereof into a power plant.
A conventional BWR typically includes conventional recirculation pumps which operate for channeling downwardly the water within the pressure vessel in a conventional annular downcomer surrounding the core, which recirculated water enters the lower plenum and flows upwardly through the core. Since the water used to generate the reactor steam also cools the reactor, systems are typically provided to ensure that adequate water is always contained within the pressure vessel and above the core during all modes of operation of the reactor, including abnormal modes such as that occurring in a conventional loss-of coolant accident (LOCA) wherein the coolant water leaks from the reactor system and must be suitably replaced for maintaining an adequate level of water within the pressure vessel above the core.
In one type of advanced BWR, a gravity-driven cooling system (GDCS) includes a pool of water located outside the pressure vessel at an elevation above the reactor core to provide makeup water in a LOCA situation for example. In order to use the GDCS makeup water, the reactor pressure vessel must be first depressurized in a conventional manner to sufficiently reduce the pressure therein so that the pressure head of the elevated GDCS makeup water is sufficient to force the makeup water into the vessel to supplant the lost reactor water for maintaining the reactor water level above the core. Since depressurization of the pressure vessel takes several minutes, the vessel continues to lose its coolant water either as a liquid or from the steam being generated and discharged therefrom, which loss of water must be suitably made up to ensure an adequate water level within the vessel.
One arrangement for ensuring adequate water level within the vessel is to provide a greater initial amount of water in the pressure vessel above the core by suitably increasing the normal elevation of the water level within the vessel. By initially providing more water within the vessel, adequate reserves of the water therein may be maintained during a LOCA situation until the vessel may be suitably depressurized and makeup water provided thereto from the GDCS pool. The increased normal water level within the vessel, however, requires a corresponding increase in the height of the pressure vessel, which correspondingly increases its manufacturing complexity and weight.
Furthermore, in another abnormal situation involving an accidental trip of all the recirculation pumps, recirculation of the coolant water within the vessel will occur solely by natural recirculation flow of the water therein with the core-heated water rising, and the relatively cooler water within the downcomer falling. By increasing the normal water level as described above. The natural recirculation flow of the coolant water within the vessel is also increased, which is effective for providing additional margin against conventionally known nuclear-thermal-hydraulic instability of the coolant water following an all-pump trip. Furthermore, the increased normal water level is also effective for improving conventional thermal margins and peak pressures for other types of plant operating transient conditions.
Analysis indicates that an increase in the normal water level within the pressure vessel of about 7 meters is required both to apply an effective gravity-driven cooling system in a LOCA situation, and to achieve suitably stable operation following an all recirculation pump trip situation for a reactor sized for generating about 1350 MWe. However, in order to provide the additional 7 meters of water above the reactor core, the entire pressure vessel must be extended 7 meters above the core which would increase the normal length thereof from about 21 meters to at least 28 meters. Such a large pressure vessel is near the current fabrication limits, and near the current crane capacity limits for assembling the vessel in the power plant. The relatively large pressure vessel increases the complexity and cost of its use within the power plant.